Disclosed herein is a method and facility to decontamination of radioactive material in connection with the processing of nuclear waste, and more particularly to the treatment of carbonaceous waste, including graphite. Nuclear reaction is often contained and stopped in the reactors with graphite sleeves. At the end of their service life, these sleeves must be treated. They then form a graphite matrix containing carbon-14 (C14).
These radioactive carbon forms must be isolated and stored in sealed containers. For this purpose, it is necessary to “break” the aforementioned graphite matrix to extract from it especially the C14 isotope. This step is usually carried out at a high temperature. Then, this isotope is precipitated for solid storage as a result of this precipitation reaction (e.g., by reaction with the lime to obtain a compound of the type CaCO3).
Conventional treatment consists of reforming by steam (or “steam reforming”) the graphite matrix, described for example in U.S. Pat. No. 6,625,248. However, the technique of said document does not really allow isolating primarily the C14 and C13 isotopes and thus ensuring acceptable radioactive waste.
There was proposed in the document FR-2943167 a very promising technique for heat treatment that allows a particularly effective isotopic separation of carbon.
Nowadays, we can achieve by such techniques an effective decontamination of tritium, carbon-14 and a part of chlorine-36. Other radionuclides are not volatile and so they are recovered in the residue at the end of a phase of steam reforming.
However, for these technologies to be of interest, it is necessary to carefully define the choice of the purge gas to obtain a maximum initial decontamination of the product associated with a loss of mass as low as possible. The term “purge gas” refers to the gas that is injected into the thermal treatment furnace (or “roaster”) during the thermal treatment step for decontamination of the graphite.
This decontamination must be sufficiently efficient without generating an excessive loss of mass in the graphite. Indeed, the graphite loss of mass generates volumes of secondary waste (carbon-14 and chlorine-36 enriched in a mineral matrix), which are expensive in terms of space for storage, the latter to be stored in deep geological formations (deep storage).
Indeed, what is called “graphite” here is a material used in reactors called NUGG for “natural uranium-graphite-gas” or MAGNOX or AVR, as the neutron flux moderator. It is actually a set of materials with sometimes marked differences in their structures, typically in their origins and, for example, their operating conditions (temperature, fluence, radiolytic corrosion, etc.), which may be different in nuclear plants, because these operating conditions have changed their structure. The variability of the reactivity of these structures is the underlying factor in the performance reproducibility.
Moreover, besides the original structural character of the graphites used in gas graphite reactors, a number of parameters affect their reactiveness and ability to be decontaminated, including:
Fluence, the effect of which is destructuring the graphite matrix: irradiated graphite is no longer graphite (in terms of crystallography) and manifests structural and nanostructural disruptions related to irradiation (as electron microscopy images can show);
Temperature, the effect of which—for temperatures of 1000° C. and above—is healing the destructuring that can be generated by high fluences: a medium-irradiated but highly heated graphite can be decontaminated less well than graphite that is very highly irradiated;
Radiolytic corrosion, which produces a scouring effect on the graphite and sparks decontamination of C-14 in the reactor;                Porosity: the pore size also plays a role.        Nanometric pores increase the probability of reaction with the purge gas, but decrease the accessibility to the active sites (where there are radioactive isotopes).        Micrometric pores decrease the probability of reaction, but increase the accessibility to the active sites.        Other influential parameters include the influence of water and the nature of the original coke.        
Given this variability of irradiated graphites and correlations between the calculations and measurements at the level of the radiological inventories that have been established, the dominant parameters for the irradiated graphite typically are:                The choice of purge gas used in the thermal treatment phase;        A large heterogeneity of the radioactivity's spatial distributions;        Radioactivity to be found mostly in the most-degraded areas, because these areas are more reactive: typically at the edge of the sheet or in the interstitial of graphite matrix.        
Thus, there remains the task to optimize the choice of the purge gas injected during the thermal treatment.